Research Papers

Methodology to Design Simulated Irradiated Fuel by Maximizing Integral Indices (ck, E, G)

[+] Author and Article Information
Jason R. Sharpe

Department of Engineering Physics,
McMaster University,
Hamilton, ON L8S 4L8, Canada
e-mail: sharpejr@mcmaster.ca

Adriaan Buijs

Department of Engineering Physics,
McMaster University,
Hamilton, ON L8S 4L8, Canada
e-mail: buijsa@mcmaster.ca

Jeremy Pencer

Computational Physics Branch, Canadian Nuclear Laboratories,
Engineering Physics,
McMaster University,
Deep River, ON K0J 1P0, Canada
e-mail: jeremy.pencer@cnl.ca

1Corresponding author.

Manuscript received May 8, 2015; final manuscript received July 8, 2015; published online February 29, 2016. Assoc. Editor: Thomas Schulenberg.

ASME J of Nuclear Rad Sci 2(2), 021017 (Feb 29, 2016) (7 pages) Paper No: NERS-15-1079; doi: 10.1115/1.4031074 History: Received May 08, 2015; Accepted July 13, 2015

Critical experiments are used for validation of reactor physics codes, in particular, to determine the biases and uncertainties in code predictions. To reflect all conditions present in operating reactors, plans for such experiments often require tests involving irradiated fuel. However, it is impractical to use actual irradiated fuel in critical experiments due to hazards associated with handling and transporting the fuel. To overcome this limitation, a simulated irradiated fuel, whose composition mimics the neutronic behavior of the truly irradiated fuel (TRUFUEL), can be used in a critical experiment. Here, we present an optimization method in which the composition of simulated irradiated fuel for the Canadian supercritical water-cooled reactor (SCWR) concept at midburnup (21.3  MWdkg1 (IHM)) is varied until the integral indices ck, E, and G are maximized between the true and simulated irradiated fuel. In the optimization, the simulated irradiated fuel composition is simplified so that only the major actinides (U233, Pu238-242, and Th232) remain, while the absorbing fission products are replaced by dysprosia and zirconia. In this method, the integral indices ck, E, and G are maximized while the buckling, k and the relative ring-averaged pin fission powers are constrained, within a certain tolerance, to their reference lattice values. Using this method, maximized integral similarity indices of ck=0.967, E=0.992, and G=0.891 have been obtained.

Copyright © 2016 by ASME
Your Session has timed out. Please sign back in to continue.


Radiation Safety Information Computational Center, 2013. https://rsicc.ornl.gov.
Rearden, B. T., Petrie, L. M., Jessee, M. A., and Williams, M. L., 2011, “SAMS: Sensitivity Analysis Module for SCALE,” Oak Ridge National Laboratory, .
Oblow, E., 1976, “Sensitivity Theory From a Differential Viewpoint,” Nucl. Sci. Eng., 59(2), p. 187. 0029-5639
Gandini, A., 1994, “A Generalized Perturbation Method for Bilinear Functionals of the Real and Adjoint Neutron Fluxes,” J. Nucl. Energy, 21(10), p. 7. 0022-3107
Jessee, M., and DeHart, M., 2011, “NEWT: A New Transport Algorithm for Two-Dimensional Discrete-Ordinates Analysis in Non-Orthogonal Geometries,” Oak Ridge National Laboratory, .
Weisbin, C. R., Marable, J. H., Lucius, J. L., Oblow, E. M., Mynatt, F. R., Peelle, R. W., and Perey, F. G., 1976, “Application of FORSS Sensitivity and Uncertainty Methodology to Fast Reactor Benchmark Analysis,” Union Carbide Corp., Oak Ridge National Laboratory, .
Williams, M., Broadhead, B., and Parks, C., 2001, “Eigenvalue Sensitivity Theory for Resonance-Shielded Cross Sections,” Nucl. Sci. Eng., 138(2), pp. 177–191. 0029-5639 10.13182/NSE00-56
Williams, M. L., 1986, “Perturbation Theory for Reactor Analysis,” CRC Handbook of Nuclear Reactors Calculations, Vol. 3, CRC Press, West Palm Beach, FL, pp. 63–188.
Dunn, M., 2000, “PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices,” Oak Ridge National Laboratory, .
Rearden, B. T., and Jessee, M., June 2011, “TSUNAMI Utility Modules,” Oak Ridge National Laboratory, .
Broadhead, B., Rearden, B., Hopper, C., Wagschal, J., and Parks, C., 2004, “Sensitivity- and Uncertainty-Based Criticality Safety Validation Techniques,” Nucl. Sci. Eng., 146(3), pp. 340–366. 0029-5639 10.13182/NSE03-2
Golouglu, S., Hopper, C. M., and Rearden, B. T., 2003, “Extended Interpretation of Sensitivity Data for Benchmark Areas of Applicability,” Proceedings of ANS 2003 Annual Meeting, the Nuclear Technology Expansion: Unlimited Opportunities, San Diego, CA, June 1–5. 0003-018X
Nava-Dominquez, A., Onder, E., Pencer, J., and Watts, D., 2013, “Canadian SCWR Bundle Optimization for the new Fuel Channel Design,” The 6th International Symposium on Superciritcal Water-Cooled Reactors, Shenzhen, Guangdong, China, Paper No. 22.
Pencer, J., Watts, D., Colton, A., Wang, X., Blomeley, L., Anghel, V., and Yue, S., 2013, “Core Neutronics for the Canadian SCWR Conceptual Design,” The 6th International Symposium on Superciritcal Water-Cooled Reactors, Shenzhen, Guangdong, China, Paper No. 21.
Pencer, J., McDonald, M., and Anghel, V., 2014, “Parameters for Transient Response Modeling for the Canadian SCWR,” The 19th Pacific Basin Nuclear Conference, Vancouver, BC, Canada, Paper No. 403.
Sharpe, J. R., Salaun, F., Hummel, D., Moghrabi, A., Nowak, M., Pencer, J., Novog, D., and Buijs, A., 2015, “A Benchmark Comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-Element Fuel Lattice Cell Parameters Using Various Computer Codes,” 35th Annual Conference of the Canadian Nuclear Society, Saint John, NB, Canada, Paper No. 33.
Greene, N., 2011, “BONAMI, Resonance Self-Shielding by the Bondarenko Method,” Oak Ridge National Laboratory, .
Petrie, L., and Rearden, B., 2011, “MCDANCOFF Data Guide,” Oak Ridge National Laboratory, .
Chapot, J., Silva, F., and Schirru, M., 1999, “A new Approach to the use of Genetic Algorithms to Solve the Pressurized Water Reactors Fuel Management Optimization Problem,” Ann. Nucl. Energy, 26(7), pp. 641–655. 0306-4549 10.1016/S0306-4549(98)00078-4
The Dakota Project, 2014. “Sandia National Laboratories,” http://dakota.sandia.gov.


Grahic Jump Location
Fig. 2

Cross-sectional view of the 64-element Canadian PT-SCWR fuel bundle concept, channel, and lattice cell

Grahic Jump Location
Fig. 1

Pu239 fission sensitivity profiles for SIMFUEL and TRUFUEL. A similarity index of GPu,fission239=0.9995 was found between these two profiles

Grahic Jump Location
Fig. 3

Methodology used to ensure the validity of comparisons. Above the dashed line indicate TRUFUEL, below the dashed line indicates SIMFUEL

Grahic Jump Location
Fig. 4

Radial and Cartesian meshing used in the one-eighth fuel cell modeled with NEWT




Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In